This invention relates to the core construction of light water nuclear reactors, and particularly to methods for minimizing the stress corrosion cracking that occurs in the metallic structural elements of the core.
While intergranular stress corrosion cracking (IGSCC) in stainless steel can be controlled or minimized in a variety of ways, the austenitic stainless steel components of light water nuclear reactors suffer a heightened susceptibility as the result of the long term irradiation which such components are subjected to in service. For example, while stainless steels in the solution or mill-annealed conditions are immune to intergranular stress corrosion cracking outside of the core, such immunity is lacking in the same materials when placed in the high irradiation field encountered inside the core.
It has been postulated that the irradiation occurring in the core exposes the alloys to IGSCC by promoting the segregation of impurities, namely phosphorus, silicon and sulfur, to the grain boundaries. Accordingly, the industry has sought to minimize this irradiation-assisted stress corrosion cracking (IASCC) by using stainless steel which is high purity in terms of these impurities. Thus, modified forms of such alloys as 348 and 304 stainless steel (using the official classification system of the American Society of Testing and Materials) in which the upper limits on each of these three elements have been lowered from the standard have been used.
These high purity grades have indeed prolonged the life of the core components, but have not completely eliminated the IASCC problem. The results have instead varied, demonstrating a lack of consistency in the alloy behavior under the irradiation conditions.
It has now been discovered that the inconsistency is virtually eliminated and an even further reduction in susceptibility to IASCC is achieved by controlling the nitrogen content of austenitic stainless steels to a maximum of 0.05 weight percent, preferably to a maximum of 0.03 weight percent. The lowered nitrogen content is combined with the other features of the alloy known to inhibit stress corrosion cracking, with the result that metallic parts of unusually long life in a nuclear reactor are achieved.